ML17285B206

Qualification of Core Physics Methods for BWR Design & Analysis

ML17285B206
Person / Time
Site:	Columbia
Issue date:	03/31/1990
From:	Larkin D, Whitcomb D WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17285B205	List: GO2-90-061, Requests Review of Encl Topical Rept WPPSS-FTS-127,Rev 0, Qualification of Core Physics Methods for BWR Design & Analysis. Upon Rept Approval,Tech Spec Changes (Per Generic Ltr 88-16) Scheduled for Jan 1992 ML17285B206
References
WPPSS-FTS-127, NUDOCS 9004250052
Download: ML17285B206 (164)

Text

QUALIFICATION OF CORE PHYSICS METHODS FOR BWR DESIGN AND ANALYSIS WPPSS-FTS-127 Mar ch 1990 Principal Engineer Bill M. Iloore Contributing Engineers Alan G ~ Gibbs James D ~

Ime 1 John D.

Teachman Duane H.

Thomsen Wi 1 liam C ~ Wolkenhauer David L ~ Whitcomb

Manager, Nuclear Fuel Date:

/ 0/

David L ~ Larkin

Manager, Engineering Analysis 5 Nuclear Fuel

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DISCLAIMER This repor t was pr epar ed by the Washington Public Power Supply System (the Supply System) for submittal to the Nuclear Regulatory Commission, NRC.

The information contained herein is accurate to the best of the Supply System's knowledge.

The use of information contained in this document by anyone other than the Supply System, or the NRC, is not authorized and with respect to any unauthorized

use, neither the Supply System nor its officers, director s,

agents, or employees assume any obligation, responsibility, or liability or makes any war r anty or representation concerning the contents of this document or its accur acy or completeness.

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ACKNOWLEDGEMENTS The Supply System acknowledges the effort expended by J.

C.

Chandler of John Elston Associates for pr epar ation of this report.

The Supply System also acknowledges R. J.

Cacciapouti of Yankee Atomic Electric Company for his review and comments on this r epor t.

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ABSTRACT This topical r epor t pr esents benchmar k analyses which demonstr ate both the validity of the Washington Public Power Supply System steady state core physics model and the gualifications of the Supply System engineer ing staf f to perf or m calculations in suppor t of the WNP-2 nuclear plant.

The main computer codes in the Supply System' steady state core physics model are the MICBURN gadolinia fuel pin depletion

code, the CASNO-2 assembly depletion code and the SIMULATE.-E three-dimensional core simulator code.

The lattice physics benchmark presented in this topical report include calculations of uniform lattice criticals and compar isons of 'local fuel pin power distr ibutions to the results of gamma scan measurements taken at Quad Cities, Unit 1.

Extensive cor e simulation benchmarks are also repor ted, based on data fr om four cycles at WNP-2, two cycles at Peach

Bottom, Unit 2,

and two Cycles of Quad

Cities, Unit 1.

The data includes hot and cold cr iticals, neutr on TIP data, and assembly gamma scan data.

The benchmark comparisons show good agreement between calculated and measured

data, ther eby demonstrating the Supply System's gualifications to set up and per form steady state cor e physics calculations for reload design and licensing applications at the WNP-2 nuclear plant.

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TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 LATTICE PHYSICS METHODS....................

10 2.1 COMPUTER CODES USED IN THE LATTICE PHYSICS ANALYSIS...

10 2.1.1 The MICBURN Burnable Absorber Depletion Program..

10 2.1.2 The CASMO-2 Lattice Depletion Program.

12 2.2 LATTICE PHYSICS BENCHMARKS.... "....:.......

14 2.2.1 Uniform Lattice Critical Benchmarks.

14 2.2.2 Quad Cities Local Power Benchmarks 16 3.0 CORE SIMULATIONMETHODS....................

30 3.1 COMPUTER CODES USED IN THE CORE ANALYSIS.........

30 3.1.1 The NORGE-B Cross-Section Representation Program 31 3.1.2 The ABLE Albedo and Boundary Leakage

- Program.

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3.1.3 The SIMULATE-E Nodal Simulator Code.

3.1.4 The FIBWR Hydraulic Model.

3.1.5 The CALTIP Instrument

Response

Model 3.2 WNP-2 BENCHMARKS.

3.3 PEACH BOTTOM BENCHMARKS 3.4 QUAD CITIES BENCHMARKS.

4.0

SUMMARY

AND CONCLUSIONS.

5.0 REFERENCES

Evaluation 32 33 35 36 37 39 42 143 146

LIST OF TABLES D

ri 2.1 2.2 2.3 3.1 3.2 3.3 3.4 3.5 3.6 HNP-2 Core Characteristics.

5 Results of Uniform Lattice Critical Analyses.

21 Summary of Local Power Benchmarks 22 Comparison of Maximum Local Peaking Factors 23 WNP-2 Cross Section Formulation 47 Summary of WNP-2 Cold Critical Predictions.

. '8 Summary of HNP-2 TIP Datasets

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49 Peach Bottom 2 Cycle 1&2 Statepoints.

50 Peach Bottom 2 Cycle 1&2 Benchmarks 51 Quad Cities 1 Statepoints 52 3.7 3.8 3.9 Quad Cities Predictions Quad Cities Predictions Quad Cities Benchmarks.

Cycle 1&2 Benchmark Summary of Cold Critical

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53 1 Cycle 1&2 Benchmark Summary of TIP

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54 Gamma Scan Benchmark Summary of Radial Power

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55 3,10 Quad Cities Benchmarks.

Gamma Scan Benchmark Summary of Axial Power

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56 3.11 Quad Cities Comparisons Gamma Scan Benchmark Axial Peak-to-Average

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LIST OF FIGURES 1.2 1.3 1.4 WNP-2 Core Arrangement.

Typical HNP-2 Power-Flow Operating Map.

Analytical Methodology for HNP-2.

Physics Methods Calculational Overview.

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9 2.1 2.2 2.3 2.4 2.5 2.6 F 1 Local Power Benchmark, Assembly CX0214, Axial Node 4.

24 Local Power Benchmark, Assembly CX0214, Axial Node 15 25 Local Power Benchmark, Assembly CX0672, Axial Node 10 26 Local Power Benchmark, Assembly GEB159, Axial Node 4.

27 Local Power Benchmark, Assembly GEB161, Axial Node 21 28

-Local Power Benchmark, Assembly GEH002, Axial Node 15 29 WNP-2 Critical Eigenvalue Calculated by SIMULATE-E.

59 3 '

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3.4 3.5 3.6 3.7 3.8 3.9 3.10 3.11 HNP-2 WNP-2 Cycle 1 TIP Predictions, 1372 MHd/MT.

Cycle 1 TIP Predictions, 1372 MHd/MT.

WNP-2 Cycle 1 TIP Predictions, 4579 MHd/MT.

WNP-2 Cycle 1 TIP Predictions, 4579 MHd/MT.

WNP-2 Cycle 2 TIP Predictions, 9297 MWd/MT.

WNP-2 Cycle 2 TIP Predictions, 9297 MHd/MT.

WNP-2 Cycle 2 TIP Predictions, 12102 MWd/MT.

WNP-2 Cycle 2 TIP Predictions, 12102 MWd/MT.

HNP-2 Cycle 3 TIP Predictions, 9903 MHd/MT.

WNP-2 Cycle 3 TIP Predictions, 9903 MWd/MT.

60 61 62 63 64 65 66 67 68 69

LIST OF FIGURES (Continued) 3.12 3.13 3.14 3.15 3.16 3.17 3.18 3.19 3.20 3.21 3.22 3.23 3.24 3.25 3.26 3.27 3.28 3.29

3. 30 3.31 3.32 3'3 3.34 3.35 WNP-2 Cycle 3 TIP Predictions, 10836 MWd/MT.

HNP-2 Cycle 3 TIP Predictions, 10836 MHd/MT.

WNP-2 Cycle 3 TIP Predictions, 11034 MHd/MT.

HNP-2 Cycle 3 TIP Predictions, 11034 MHd/MT.

HNP-2 Cycle 3 TIP Predictions, 12245 MHd/MT.

HNP-2 Cycle 3 TIP Predictions, 12245 MHd/MT.

HNP-2 Cycle 3 TIP Predictions, 12903 MHd/MT.

WNP-2 Cycle 3 TIP Predictions, 12903 MWd/MT.

WNP-2 Cycle 3 TIP Predictions, 13974 MHd/MT.

WNP-2 Cycle 3 TIP Predictions, 13974 MHd/MT.

WNP-2 Cycle 4 TIP Predictions, 11245 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 11245 MWd/MT.

HNP-2 Cycle 4 TIP Predictions, 12903 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 12903 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 13563 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 13563 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 14428 MHd/MT.

WNP-2 Cycle 4 TIP Predictions, 14428 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 15473 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 15473 MHd/MT.

WNP-2 Cycle 4 TIP Predictions, 16531 MHd/MT.

HNP-2 Cycle 4 TIP Predictions, 16531 MHd/MT.

Peach Bottom Unit 2 Critical Eigenvalue Calculated SIMULATE-E Summary of TIP Predictions Peach Bottom Benchmark.

by 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93

LIST OF FIGURES (Continued)

D ri Peach Bottom 2 Cycle 1 TIP Predictions, 1113 MHd/MT.

94 Peach Bottom 2 Cycle 1 TIP Predictions, 1113 NWd/MT.

95 Peach Bottom 2 Cycle 1 TIP Predictions, 2816 MHd/MT.

96 Peach Bottom 2 Cycle 1 TIP Predictions, 2816 MHd/NT 97 Peach Bottom 2 Cycle 1 TIP Predictions, 5178 MHd/MT.

98 Peach Bottom 2 Cycle 1 TIP Predictions, 5178 MHd/MT.

99 Peach Bottom 2 Cycle 1 TIP Predictions, 5800 MWd/MT.

100 Peach Bottom 2 Cycle 1 TIP Predictions, 5800 MWd/MT 101 Peach Bottom 2 Cycle 1 TIP Predictions, 6731 MHd/MT.

102 Peach Bottom 2 Cycle 1 TIP Predictions, 6731 MHd/MT.

103 Peach Bottom 2 Cycle 1 TIP Predictions, 11133 MHd/MT.

104 Peach Bottom 2 Cycle 1 TIP Predictions, 11133 MHd/MT.

105 Peach Bottom 2 Cycle 2 TIP Predictions, 10726 MHd/MT.

106 Peach Bottom 2 Cycle 2 TIP Predictions, 10726 MHd/NT.

107 Peach Bottom 2 Cycle 2 TIP Predictions, 11078 MWd/MT.

108 Peach Bottom 2 Cycle 2 TIP Predictions, 11078 MHd/MT.

109 Peach Bottom 2 Cycle 2 TIP Predictions, 12754 MHd/MT.

110 Peach Bottom 2 Cycle 2 TIP Predictions, 12754 MWd/MT.

111 Peach Bottom 2 Cycle 2 TIP Predictions, 13812 NHd/MT.

112 Peach Bottom 2 Cycle 2 TIP Predictions, 13812 MWd/NT.

113 Quad Cities Critical Eigenvalue Calculated by SIMULATE-E.

114 Summary of TIP Predictions Quad Cities Benchmark, 115 Quad Cities 1 Cycle 1 TIP Predictions, 712 MWd/MT 116 Quad Cities 1 Cycle 1 TIP Predictions, 712 MWd/MT....

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LIST OF FIGURES (Continued) 3.60 3.61 Quad Cities 1 Cycle 1 TIP Predictions, 2239 MHd/MT; Quad Cities 1 Cycle 1 TIP Predictions, 2239 MHd/MT.

118 119

3. 62 Quad Cities 1 Cycle 1 TIP Predictions, 4737 MHd/MT.

120 3.63 3.64 3.65 3.66 3.67 3.68 3'9 3.70 3.71 3.72 3.73 3.74 3.75 Quad Cities 1 Cycle 1 TIP Predictions, 4737 MHd/MT.

Quad Cities 1 Cycle 1 TIP Predictions, 6807 MHd/MT.

Quad Cities 1 Cycle 1 TIP Predictions, 6807 MHd/MT.

Quad Cities 1 Cycle 1 TIP Predictions, 8068 MHd/NT.

Quad Cities 1 Cycle 1 TIP Predictions, 8068 MHd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 7964 MHd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 7964 NHd/NT.

Quad Cities 1 Cycle 2 TIP Predictions, 9141 MWd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 9141 MHd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 13198 MHd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 13198 MHd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 13741 MHd/MT.

Quad Cities 1 Cycle 2 TIP Predictions, 13741 MHd/MT.

121 122 123 124 125 126 127 128 129 130 131 132 133 3.76 3.77 3.78 3.79 Quad Cities Radial Power Benchmark Assembl Activity Levels Quad Cities Radial Power Benchmark Radial Distribution at Axial Plane 7

Quad Cities Radial Power Benchmark Radial Distribution at Axial Plane 15.

Quad Cities Radial Power Benchmark Radial Distribution at Axial Plane 18.

y Averaged La-140 La-140 La-140 134

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LIST OF FIGURES (Continued)

Core Average Axial La-140 Distribution Quad Cities Gamma Scan Benchmark.

138 Axial La-140 Di stribution Gamma Scan Benchmark.

Axial La-140 Distribution Gamma Scan Benchmark.

Axial La-140 Distribution Gamma Scan Benchmark.

Axial La-140 Distribution Gamma Scan Benchmark.

for Assembly CX0553 for Assembly GEH023 for Assembly CX0214 for Assembly GEB161 Quad Cities 139 Quad Cities 140 Quad Cities

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1.0 INTRODUCTION

The Washington Public Power Supply System ("Supply System" ) operates the WNP-2 nuclear plant near Richland, Washington.

WNP-2 is a BHR/5 installa-tion using a nuclear steam supply system designed by the General Electric Company (GE),

who also provided the initial core fuel for the reactor.

Reload nuclear fuel and associated analyses are currently provided by Advanced Nuclear Fuels Corporation (ANF).

Nuclear fuel design and assem-bly design criteria are provided by the fuel vendors.

The core character-istics and core arrangement of HNP-2 are given in Table 1.1 and Figure 1.1, respectively.

A typical HNP-2 power-flow operating map is shown in Figure 1.2.

A desire to understand plant performance under conditions both within and outside the normal operating envelope has influenced a

number of licen-'ees to develop in-house analytical methods for predicting normal plant operation and responses to off-nominal event initiators.

The potential for enhanced cost effectiveness, more rapid analytical greater latitude in evaluation of al ternati ves relative dependence on outside agencies has also 'contributed to the licensees to develop in-house analytical capabilities.

response, and to analytical incentive for While specific methods may vary from licensee to licensee, the overall approach to the problem invariably separates the analytical responsibi 1-i ty at the core boundary.

Reactor physics analysis provides a detailed, steady-state prediction of neutron flux effects using system performance

as a boundary condition.

Systems analysis provides a detailed, transient prediction of plant response based on a simplified core transient model.

Inputs for the simplified core transient model used in the systems analy-sis are derived from the reactor physics analysis.

The Supply System has developed an overall analytical methodology to sup-port the NNP-2 nuclear plant.

A simplified overall calculational flow used in this methodology is shown in Figure 1.3.

This document covers the segment labeled "Steady State Core Physics Analysis;"

the remaining segments are reported separately.

The Supply System's steady state core physics methods are based on the Electric Power Research Institute (EPRI) code package depicted in the flowchart in Figure 1.4.

The main computer codes are the CASMO-2 fuel bundle lattice 'hysics depletion code and the

'IMULATE-E three-dimensional core simulation code.

Both of these codes represent current technology for reactor analysis and are described further in later sec-tions of this report.

MICBURN provides a detailed representation of the depletion of a

single gadolinia-bearing fuel rod; NORGE-B provides a

nuclear cross-section data link from CASMO-2 into SIMULATE-E.

CALTIP pro-vides incore instrument responses,

plotting, and statistical evaluation.

The Supply System uses these computer programs and associated methodolo-gies for plant operations support applications (such as core follow analy-

ses, development of target control rod patterns, predictions of startup critical rod patterns, and operating strategy evaluations),

independent

design'erification calculations,'eload fuel/core design

analyses, and safety analyses.

The steady state core physics methods described in this report are also used to develop the necessary neutronics data input to the Supply System's transient analyses.

This report describes the steady state core physics methods used by the Supply System for BWR core analysis and provides qualification of the analytical methodologies that will be used to perform safety-related analyses in support of the WNP-2 operating license.

Applications of this methodology are reported separately, as are systems analysis methods and models.

The work reported herein is intended to satisfy the guidelines of Generic Letter 83-111, which established that user qualifications are subject to regulatory review consistent with the technology used by the licensee'.

The ma)or analytical tools used in the Supply System methodology have been used by other BWR licensees and have been reviewed in similar applications The qualification of the Supply System's steady state core. physics meth-ods is based on comparisons of calculated core parameters to measured data from WNP-2, Peach Bottom Unit 2, and Quad Cities Unit 1.

All of the model preparation and benchmarking calculations represent work performed by the Supply System.

The computer codes and the calculations supporting this work are documented,

reviewed, and controlled by formal procedures which are encompassed within the Supply System's nuclear quality assur-ance program.

Subsequent chapters of this report cover the one-and two-dimehsional analysis of fuel assembly lattices and the three-dimensional simulation of core-wide phenomena.

In each of these

chapters, the text describes the computer models and methodologies used for these analyses.

Detailed benchmarks provided in both sections show consistently accurate predic-tions of published and operational data for Quad Cities Unit 1, Peach Bot-tom Unit 2, and NNP-2.

TABLE 1.1 WNP-2 CORE CHARACTERISTICS Reactor Type/Configuration Rated Core Power:

Rated Core Flow:

Reactor Pressure at Rated Conditions:

Number of Fuel Assemblies:

Number of Control Rods:

Number of Traversing In-Core Probe Locations:

BWR/5 2-loop jet pump recirculation system 3,323 HW thermal 108.5x10 ibm/hr 1035 psia 764 185 43

Figure 1.1 WNP-2 Core Arrangement 60 58 56 54 52 50 48 46 44 42 4038 3634 32 30 28 26 24 22 20 +

16 14 12 10 08 06 04 02

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g 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59

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TIP Locations

'ontrol Rod Locations

-6" 890365.

100 90 8070'NP-2 Two-Pump Operation Line FGV No.

PoO.

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15Hz 2

100 3

0 60 Hz 10 104% Xe Rod Lin I

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0% Xe Rod Line 106 Percent Power I

oy 60 40 13

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30 20 10 0

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. 25 50 Percent Core Flow%

75 100 890365.2 Rev. Feb. 1990

Figure 1.3 Analytical Methodology for MfNP-2

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Figure IA Physics Methods Calculational Overview

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(Gadollnia Characteristics)

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(Cross Sections)

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(instrument Factors)

Steady4tate Physics Transient Analysis

2.0 LATTICE PHYSICS METHODS The detailed application of reactor theory is contained in the lattice physics analysis.

Direct prediction of neutron and nuclide performance within the assembly lattice is accomplished with neutron transport theory and translated into an output form usable by the core analysis programs.

2.1 COMPUTER CODES USED IN THE LATTICE PHYSICS ANALYSIS This section describes the computer programs used by the Supply System in the analysis of detailed fuel performance characteris-tics.

Each computer code included in this methodology package is sub)ect to qualification in accordance with the Supply System's engi-neering quality assurance

program, which includes verification and validation requirements for computer codes used in design and safety analysis applications.

2.1.1 The MICBURN Burnable Absorber Depletion Program The homogeneous treatment of the fuel pellet used in CASMO-2 is acceptably accurate for those fuel rods in the core that contain no burnable absorber material.

However, BNR fuel assemblies usually contain a

number of fuel rods seeded with gadolinia burnable absorber.

The gadolinia fuel rods are characterized with the MICBURN computer code MICBURN provides a one-dimensional neutron trans-4 port analysis of the fuel rod and models the consumption of gadolinia in annular rings as burnup accumulates.

The Supply System's methodology for MICBURN uses the same 25-group structure cross-sect1on 11brary as is used for CASMO-2.

The gadolinia fueled region is modeled in MICBURN with ten radial rings to provide sufficient detail to calculate the radial flux distribution.

These radial rings are expanded to twenty radial regions for the actual gadolinia deplet1on.

The cladding region is modeled w1th one radial ring, the moderator is modeled with two radial rings and the buffer reg1on w1th two rad1al rings.

The buffer region's composition is made up of the eight fuel rods and moderator adjacent to the gadol1nia fuel rod.

The enrichment of the fuel in the buffer region is chosen to correspond to the enrichment of the fuel rod enrichment type of which there is the largest number 1n the lattice.

The results of,the MICBURN analysis are used as cross sec-t1on informat1on by CASH0-2, wh1ch in turn assimilates the individual fuel rod performance characteristics into lat-tice performance factors for use in the core analysis.

As products of the same development

agency, MICBURN and CASMO-2 use the same cross sect1on libraries,'hich are based on a modified version of ENDF/B-III.

The CASHO-2 Lattice Depletion Program Detailed lattice physics analyses for WNP-2 are performed using CASMO-2 This code uses the method of transmission probabilities to solve the neutron transport equation within a two-dimensional representation of a fuel assembly.

The CASMO-2 analysis covers analytical segments which were treated separately in earlier methodologies.

Because of the automated link among the segments performed by CASH0-2, no explicit resonance, pin cell, or neutron spec-trum calculations are required.

From input data which pro-vide a physical description of the fuel rods, the

bundle, and the adjacent control rod, CASHO-2 performs in sequence the resonance calculation, a pin cell calculation for. each specific fuel rod type in the

bundle, a

one-dimensional spectrum calculation, and a two-dimensional power distribu-tion and k-infinity calculation.

The calculational flow and code input requirements are described in Reference 5.

The Supply System's methodology uses the CASHO-2 25-group energy group cross-section library described in Reference 5.

The Supply System uses the defaults specified in CASHO-2 for condensing the 25-group structure down to the 17 energy groups for the one-dimensional pin cell calcula-I tion and 7

energy groups for the two-dimen'sional lattice physics calculation.

Specific input provided to the CASMO-2 calculation includes composition data for the fuel and structure, burnable absorber characteristics calcu-lated by

MICBURN, assembly and control dimensions, and internal fuel rod arrangement.

CASMO-2 output is retained on magnetic storage for manipulation by NORGE-B into SIMULATE-E compatible input.

In the original qualification of CPM EPRI provided a

number of predictions of uniform lattice criticals pub-lished in the open literature.

To help qualify the k-effective determined by the Supply System using CASM0-2, comparisons were made between CASMO-2 analyses and a

number of these uniform lattice criticals.

The results of the Supply System's calculations of the uniform lattice critical experiments are described in Section 2.2.1.

To qualify the local power distribution determined by the Supply System using CASM0-2, comparisons were made between the calculated pin powers from CASMO-2 and gamma scan results from EPRI-sponsored testing at the end of second cycle operation at Quad Cities Unit 1.

The results of these local power comparisons are given in Section 2.2.2.

2.2 LATTICE PHYSICS BENCHMARKS E

Detailed analysis of published data was undertaken to provide assur-ance that the lattice physics analyses performed in support of the WNP-2 core simulation work will provide quality results.

Two aspects of the lattice physics analysis were selected for benchmark-ing:

the calculation of k-effective and the prediction of local power distributions.

2.2.1 Uniform Lattice Critical Benchmarks Consi stent with the qualification performed by EPRI for

CPM, the Supply System selected fifteen critical experi-ments from the TRX and ESADA published data for qualification of CASMO-2.

These cases comprised eight U02 configurations from the TRX series and seven mixed oxide configurations from the ESADA series.

The TRX criticals had fuel pellet densities ranging from 7.52 gm/cm to 10.53 gm/cm and all eight had rod enrich-ments of 1.3 weight percent U-235.

The ESADA criticals all had a compacted powder fuel of density 9.54 gm/cm con-sisting of 2.0 weight percent Pu02 in natural U02.

In five of the ESADA experiments the plutonium was nominally 8/

Pu-240; in the remaining two it was nominally 24K Pu-240.

The two groups of experiments span a

range of lattice pitches with the equivalent cylindrical cell radii varying from 0.818 cm (TRX-4 and -6) to 1.978 cm (ESADA-7).

The corresponding cell radius for NNP-2 is 0.917 cm.

All of the selected criticals were modeled by CASMO-2 as single pin cells, with core leakage accounted for in a fun-damental mode approximation using the experimental geomet-ric buckling.

The smaller pitch criticals were modeled as three-region, homogenized cells.

In the larger pitch crit-

icals, the moderator region was divided into two subre-

gions, and Gaussian integration points were inserted in each of the four resulting regions.

For the smallest pitch cri ticals,

. which were most typical of HNP-2 geome-try, this refinement proved to be unnecessary, exhibiting a negligible'effect on the calculated k-effective.

The TRX critical experiments used Lucite spacers.

Because Lucite is neutronically similar to the water moderator, no adjustment was necessary to account for the presence of spacers.

The ESADA criticals used aluminum spacers, which necessitated some adjustment in the two-dimensional CASMO-2 analysis.

ESADA spacer worths were estimated by running CASMO-2 with and without axially homogenized spac-ers.

A two-group perturbation theory analysis was used to correct the axially homogenized spacer worth for the effects of spacer localization.

The corrected spacer worths vary from 6.84 mk for the small-pitch ESADA-1'exper-iment to 0.11 mk for the large-pitch ESADA-7 experiment.

The results of the uniform lattice critical analyses with CASHO-2 are shown in Table 2.1.

The eight TRX urania crit-icals have a

mean k-effective of 0.99616 "with a standard deviation of 0.00184.

The seven ESADA mixed oxide criti-cals have a

mean k-effective of 1.00872 with a

standard deviation of 0.00782.

In the ESADA

series, the k-effective values are generally higher for the larger pitch criticals while for the smaller pitch criticals, which are similar to HNP-2, k-effective values are in the same range as for the TRX series.

2.2.2 Quad Cities Local Power Benchmarks Gamma scan data for fuel irradiated during the first two operating cycles at Quad Cities 1

were published by EPRI Pin-by-pin gamma scans were performed on six assemblies as a

part of the EPRI study.

The three U02 assemblies (GEH002,

CX0672, and CX0214) and two mixed oxide assem-blies (GEB159 and GEB161) were used for the Supply System benchmark.

The first two of these assemblies were located about halfway between the core edge and the core center; the remaining three were located near the core center.

Detailed core modeling based on the first two operating cycles of Quad Cities 1

was undertaken to demonstrate the adequacy of the Supply System's methods.

Core design information as published by EPRI was used to model and deplete the core to the end of Cycle 2.

The modeling of the depletion of the first two cycles and prediction of nodal power distribution results are described in Chap-ter 3.0 of this report.

Barium-140 inventories at the end of Cycle 2

were predicted explicitly in both the lattice calculations and the core calculations for comparison against the measured La-140 activity levels.

The predic-tions of Ba-140 were based on detailed power shapes from the last six statepoi nts

( 125 days) of Cycle 2,

using a

piecewise linear representation of the power history for this time.

Contributions to 'the end of Cycle 2

Ba-140 inventory from earlier irradiation times were shown to be negligible.

The local power benchmarks required an additional analyti-cal step beyond the normal core analysis.

In addition to the lattice physics calculations and the three-dimensional depletions, the benchmark also required a return to the lattice physics analysis for detailed fuel performance cha-racteristics.

The SIMULATE-E analysis provided void his-

tory, instantaneous

void, and nodal exposure values for each node in the core.

These values were used to interpo-late among the CASMO-2 cases for the lattice types in question to determine local power distributions,'hich were then used in the explicit determination of Ba-140 in-ventories for comparison against the experimental data.

Since the purpose of this benchmark is to evaluate the lo-cal power distribution, the measured and calculated re-sults were normalized to one on each plane of each pin-scanned assembly.

This procedure eliminates uncertainties associated with the SIMULATE-E nodal power distribution.

The nodal power uncertainties are evaluated separately by comparisons to TIPs and assembly gamma scans in Chapter 3.0.

In this

report, the accuracies of predictions are expressed in terms of various statistical measures.

These measures are defined either in terms of the fractional cal-culational error

(

Ci - Illi )

x 1001.

(1) where:

ei fractional calculational error in percent ci normalized calculation, and mi - normalized measurement.

or in terms of the difference, ci - mi (2) where di the calculation difference.

The mean values of these quantities are defined by Eei N

(3) and Ed, m

(4)

N

Here, N denotes the total number of points in the sample.

Often, calculated and measured results are normalized to a

common value; in such cases, dm is zero.

The degree of variation of ei and di about their mean

. values can be characterized by their.standard deviations, expressed as a percent of the mean measured value:

E(ei em)2 creV.)

( - )

'-1 (5) and E(di dm) 2

'001.

i/x.

edV)

(' - )~

x N-1 (6)

Here, M denotes the mean measured result.

Alternatively, the accuracy of a prediction can be cha-racterized by the root mean square values of ei or di, expressed as a percentage of M:

Eei 2 Erms (1o)

~ (

)

N Ed i2, 100/

D (1) =( )~x N

H (7)

(8)

Results of the local power gamma scan benchmarks ar'e sum-marized in Table 2.2.

There, the center of axial node N

is located (6N 3) inches above the bottom of the active fuel.

These results show a core-wide standard deviation, vd, of 3.20/.

based on the differences between measured and predicted values and including all data points.

If the mixed oxide bundles are excluded from the statistics, this standard deviation becomes 3.11/.

Measurement error for this database was published at 1.7'/;

these results indi-cate a calculational error approximately the same as the measurement error.

The measured and calculated maximum local peaking factors are compared in Table 2.3.

This comparison is important because monitored values for linear heat generation rate (LHGR) and critical power ratio (MCPR) are determined by the maximum local power peaking factor in any given node.

Comparison of measurement and calculation for maximum local peaking independent of position within the lattice shows an standard deviation, ee, of 2.14'/.

The measure-ment error was published at 1.7'/ for this segment of the database as well.

Selected local power comparisons are given in Figures 2.1 through 2.6.

These comparisons illustrate the range of calculations and are typical of the datasets which are not reproduced here.

TABLE 2.1 RESULTS OF UNIFORM LATTICE CRITICAL ANALYSES Experiment TRX-1 TRX-2 TRX-3 TRX-4 TRX-5 TRX-6 TRX-7 TRX-8 CASMO-2 RESULTS FOR TRX CRITICALS Experimental 28.37 30

~ 17 29.06 25.28 25.21 32.59 35.47 34.22 CASMO-2 v

0.99421 0.99749 0.99733 0.99404 0.99361 0.99785 0.99753 0.99722 Average K-effective:

Standard Deviation:

0.99616 0.00184 Experiment ESADA-1 ESADA-3 ESADA-4 ESADA-6 ESADA-7 ESADA-12 ESADA-13 CASMO-2 RESULTS FOR ESADA CRITICALS Experimental B

li

69. 6 90.0 104.72 98.4 50.3 79.5 73.3 CASMO-2 0.99809 0,99816 1.01647 1.01715 1.00901 1.01109 1.01105 Average K-effective:

Standard Deviation:

1.00872 0.00782 Buckling expressed in m

CASMO-2 calculated k-effectives for the ESADA criticals have been adjusted to account for spacer worth.

TABLE 2.2 SUNMARY OF LOCAL PONER BENCHHARKS Assembly CX0214 CX0672 GEB159 GEB161 GEH002 Axial Location Qhdr2 3

4 9

10

-15 16 21 22 3

4 9

10 15 16 21 22 3

4 1'0 15 16 21 22 3

4 9

10 15 16 21 22 3

4 9

10 15 16 21 22 Nodal Relative Hater Density 0.022 0.052 0.302 0.350 0.527 0.551 0.635 0.645

0. 018 0:044 0.282 0.335 0.520 0.545 0.631 0.642 0.039 0.087 0.398 0.443 0.597 0.617 0.689 0.698 0.039 0.087 0.398 0.443 0.597

.0. 617

0. 689 0.698 0.040 0.089 0.399 0.443 0.592 0.612 0.684 0.692 Nodal Exposure QKD~3 16.117 17.642 20.270 20.248 19.747 19.402 15.148 12.906 15.768 17.352 20.089 19.980 19.501 19.127 15.383 13.256 10.878 11.520 10.506 10;435 10.020 9.964 7.854 6;552 10.878 11.520 10.506 10.435 10.020 9.964 7.854 6.552 10.523 11.138 10.215 10.061 9.531 9.434 7.334 6.175 Standard Deviation 3.32 4.34 2.77 3.29 3.09 3,90 3.51 3.08 4.23 4.21 3.05 3.28 2.99 3.04 2.89 3.21 3.33 3.69 3.67 3.51 2.92 3.04 3.26 3.31 4.06 3.41 3.82 2.34 3.06 3.26 4.79 4.72 2.32 2.36 2.63 2.77 3.04 3.13 2.27 2.40 TABLE 2.3 COMPARISON Of MAXIMUM LOCAL PEAKING fACTORS Assembly Idm~

CX0214 CX0672 GEB159 GEB161 GEH002 Ax)al 3

4 9

10 15 16 21 22 3

4 9

10 15 16 21 22 3

4 9

10 15 16 21 22 3

4 9

10 15 16 21 22 3

4 9

10 15 16 21 22 Measured 1.1082 1.0785 1.0910 1.0657 1.1262 1.1232 1.1314 1.1142 1.1057 1.0801 1.0965 1.0983 1.0956 1.0706 1.0885 1.1013

1. 1499 1.1153 1.1156 1.0994

'.1030 1.1024 1.1335 1.1593 1.1147 1.1102 1.1071 1.0887 1.0944 1.1099 1.1312 1.1720 1.1031 1.0986 1.1103 1.1002 1.1180 1.1187 1.1351 1.1351 Calculated 1.0873 1.0795 1.0751 1.0729 1.0792 1,0780 1.0954 1.1032 1.0869 1.0821 1.0701 1.0727 1.0750 1.0733 1.0975 1.1046 1.1904 1.1836 1,1516 1.1397 1.1123 1.1121 1.1562 1.1777 1.0960 1.0785 1.0845 1.0847 1.1124 1.1156 1.1462 1.1601 1.1174 1.1146 1.1168 1.1120 1.1159 1.1224 1.1369 1.1405 Error a)Q2

-1.892 0.096

-1.459 0.679

-4.172

-4.017

-3.175

-0.985

-1. 701 0.186

-2.413

-2.330

-1.876 0.250 0.833 0.302 3.520 6.122 3.225 3.666 0.846 0.876 2.000 1.584

-1.679

-2.860

-2.048

-0.369 1.651 0.510 1.325

-1.022

1. 291 1.455 0.584 1.074

-0.188 0.328 0,156 0,478 Figure 2.1 Local Power Benchmark, Assembly CX0214, Axial Node 4

1.0671 1.0132 0.9800 0.9311

-8.1596

-8.0973 1.0293 0.9738

-5.3874 1.0329 0.9932

-3.8469 0.9024 0.9609 6.4809 0.9681 1.0056 0.9311 0.9601

-3.8178

-4.5238 0.9230 0.9295 0.7028 1.0702 1.0590 1.0353 1.0412

-3.2607

-1.6794 1.0432 1.0355

-0.7455 0.9212 0.9753 5.8807 0.9126 0.9295 1.8521 0.9727 '.9747 0,9816 0.9905 0.9174 1.6169 0.9991 1.0306 3.1576 1.0051 1.0785 0.9972 0.9738 1.0353 0.9816

-3.1081

-4.0063

-1.5632 0.9976 0.9954

-0.2200 1.0588 1.0617 0.2733 1.0678 1.0412

-2.4961

1. 0021 0.9905

-1.1605 0.9702 0.9728 0.2759 0,9747 0.9766 0.1905 1.0006 1.0165 1.5858 1.0636 1.0450 0.9932 1.0355

-6.6264

-0.9111 0.9859 1.0306 4.5403 1.0104 0.9123 0.9954 1.0165

-1.4801 11.4211 0.9502 1.0410 9.5550 1.0435 1.0795 3.4546 0.9417 0.9609 2.0462 0 '627 0.9753 1.3169

1. 0479 1.0617 1.3220 1.0495 1.0795 2.8577 0.9407 1.0062 6.9647 Meas.

Calc.

I frr f ms' 4.36%

Standard Deviation, cd(%)

4.34%

Figure 2.2 Local Power Benchmark, Assembly CX0214, Axial Node 15

1. 1262

1. 0304 1.0636 0.9939

-5.5581

-3,5385 1.0563 1.0149

-3.9142 1,0374 1.0354

-0.1847 1.0228 1.0325, 0.9495 1.0046 0.9939

-1.0603 0.9707 0.9793 0.8878 0.9344 0.9433 0.9535 1.0842 1.0511 1.0223 1.0259

-5.7117

-2.3998 1.0219 1.0207

-0.1164 0.9705 1.0085 3.9126 0,9423 1.0140 0.9433 0.9615 0.1083

-5.1827 0.9908 0.9751 0.9594 0,9631

-3.1719

-1.2261 1.0654 1.0568 1.0149 1.0223

-4.7398

-3.2669 0.9546 0.9594 0.4977 0.9019 0.9369 3.8886 0.9520 0.9493

-0.2844 1.0079 1.0553 4.7074 1.0446 1.0259

-1.7889 0.9584 0.9631 0.4932 0.9107 0.9369 2.8745 0,8931 0.9385 5.0920 0.9300 0.9807 5.4591 1.0066 1.0354 2.8686 1.0203 1.0207 0.0386 0,9693 1.0038 3.5590 0.9621 0.9493

-1.3333 0.9664 0.9807 1.4845 0.9731 1.0106 3.8536 1.0618 1.0792 1.6432 1.0206 1.0325 1.1628 0.9794 1.0085 2.9626 1.0444 1.0553 1.0421 1.0882 1.0792

-0.8257 Meas.

Calc.

'/ Err E

ms 3.02K Standard Deviation, ad(%)

3.091.

+-

+

Figure 2.3 Local Power Benchmark, Assembly CX0672, Axial Node 10 1.0338 1.0236

-0.9870 0.9555 0.9658 1.0806 0.9913 0.9927 0.1455 1.0099 1.0107 0.0780 0.9522 0.9989 4.8961 0.9412 0.9658 2.6167

0. 9714 0.9698

-0.1660 0.9209 0.9395 2.0176 1.0661 1.0264

-3.7317 1.0236 1.0314 0.7563 1.0510 1.0232

-2.6441 0.9518 0.9913 4.1438 0.9094 0.9395 3.3150 1.0304 0.9933

-3.5997 1.0102 0.9723

-3.7545 1.0040 0.9788

-2.5140 1.0339 1.0165

-1.6869 1.0284 0.9927

-3 '755 1.0755 1.0264

-4.5730 0.9929 0.9723

-2.0776 0.9785 0.9573

-2.1669 1.0089 0.9715

-3.7068 1.0116 1.0541 4.2050 1.0749 1.0314

-4.0513 0,9688 0,9788 1.0303 0.9733 0.9573

-1.6425 0.9284 0.9603 3.4345 0.9876 0.9993 1.1837 1.0147 1.0107

-0.3899 1.0983 1.0232

-6.8405 1.0379 1.0165

-2.0605 0.9831 0.9715

-1.1817 0.9570 0.9993 4.4142 1.0133 1.0241 1.0683 1.0482 1.0727 2.3356 1.0080 0.9989

-0.9095 0.9709 0.9913 2.0982 0.9935 1.0541 6.1023 1.0097 1.0727 6.2354 0 '797 1.0244 4.5646 Meas.

Calc.

% Err E

ms 3.20%

Standard Deviation, crd(%)

3.28%

Figure 2.4 Local Power Benchmark, Assembly GEB159, Axial Node 4

1.0004 0.9783

-2.2123

'1.0015 0,9882

-1.3203 0.9467 0.9794 3.4558

1. 0074 1.0073

-0.0048

1. 0154 1.0089

-0.6422 0.9919 0.9882

-0.3667 0.9557 0.9321

-2.4703

1. 0165

1. 1144

0. 9839

1. 0619

-3.2030

-4.7150 1.0949 1.0097 1.0614 0.9884

-3.0585

-2.1063 0.9654 0.9649

-0.0491 1.0327 0.9839

-4.7247 0.9140 0.9549 0.8394 0.9412

-8.1598

-1.4427 0.9569 0.8894 0.9625

'.8342 0.5883

-6.2111 0.9794 0.9794 0.0047 1,1003 1.0619

-3.4893 0.9515 0.9412

-1,0887 1.0359 1.0568 2.0237 1.0872 1.1207 3.0758 1.0886 1.0784

-0.9386 1.0782 1.0614

-1.5546 0.9421 0.9625 2.1715 1.0309 1.0568 2 '179 0.6118 0.6375 4.1981 1.1038 1.1836 7.2272 1.0017 1.0073 0,5583 1.0018 0,9884

-1.3291 0.9014 0.8342

-7.4575 1.0812 1.1207 3.6451 1.1153 0.8389 1.1836 0.8100 6.1224

-3.4514 1.0728 1,1346 5.7611 0.9859 1.0089 2.3350 0.9350 0.9649 3.1997 1

~ 0598 1.0784 1.7567 1.1043 1.1346 2.7401 1.0245 1.0901 6.3970 Neas.

Calc,

% Err E

ms 3.65%

Standard Deviation, cad(%)

3.69%

Figure 2.5 Local Power Benchmark, Assembly GEB161, Axial Node 21 1.0507 1.0876 3.5145

1. 1312 1.1462

1. 3252 1.0548 1.0444

-0.9875 1,0848 1.1084 2.1796 0.9537 0.9685 1.5492 0.9512 0.9724 2.2260 0.9767 1.0058 2.9707 1.0155 1.0058

-0.9628 0.7813 0.6609

-15.402 Meas.

Calc.

% Err E

ms 5.52%

Standard Deviation, ad(%)

4.79%

Figure 2.6 Local Power Benchmark, Assembly GEH002, Axial Node 15 1.061

1. 100 3.658 1.099 1.096

-0.249 1.072 1,061

-0.991 1.071 1.051

-1.814 1.094 1.098 0.323 1.111 1.116 0.407 1,091 1.096 0.478 1.027 1.008

-1.818 1.118 1.064 1.067 1.019

-4.527

-4.201 1.045 1.043 1.005 1.019

-3.919

-2.318 1.086 1.066

-1.773

1. 020 1.043 2.304 1.092 1.067

-2.282 1.003 0.951 0.947 0.925

-S.573

-2.718 0.937 0.924

-1.389 0.990 0.948

-4.252 1.066 1.061

-0.464 1.053 1.019

-3.253 0.908 0.894

-1.584

0. 912

0. 914 0.902 0.891

-1.160

-2.475 0.936 0.927

-0.999 1.057 1.050

-0.626 1.032 1.051 1.894 1.004 1.005 0.061 0.913 0.919 0.911 0.902

-0.137

-1:870 0.906 0:899

-0.760

0. 910 0.913 0.304 1.011 1.035 2.467

1. 051

1. 019

-3.079 0.948 0.902 0.924 0.891

-2.520

-1.235 0.909 0.899

-1.113 0.872 0.891 2.155 0.908 0.926 2.015 1.089 1.098 0.825 1.064 1.066 0.205 0.961 0 '48

-1.313 0.910 0.927 1.830

+

0.877 0.913 4.016 0.902 0.926 2 '63

0. 928

0. 947

1. 979

1. 040

1. 100 5.828 1.091 1.116 2.311 0.999 1.043 4.391 1.040 1.050 0,987 0.997 1.035 3.866 1.039 1.100 5.924 0.956 1.061 10.999 Meas.

Calc.

% Err E

ms 3,01%

Standard Deviation, crd(%)

3.04%

3.0 CORE SIMULATION METHODS The detailed performance characteristics of the fuel lattice are col-lapsed into nodal performance indicators which are linked in the nodal simulator code.

These performance indicators, which take the form of regional macroscopic cross

sections, are coupled with modified coarse-mesh diffusion theory (MCHDT) to provide an overall flux and power shape for the reactor core.

The core analysis in turn provides nodal performance indices which can be keyed back to the lattice physics analysis to determine detailed fuel performance characteristics at any point in the operating cycle.

This chapter covers the computer codes and methodologies used for repre-sentation of cross

sections, the calculation of albedo

factors, the three-dimensional core simulation; and the calculation of neutron TIP responses.

3.1 COMPUTER CODES USED IN THE CORE ANALYSIS The computer codes used by the Supply System in modeling the three-dimensional behavior of a BWR's core are shown in Figure 1.4.

The codes HICBURN and CASMO-2 have already been described in Section 2.0.

This section will describe the remaining codes used in the steady state modeling.

These codes are NORGE-B,

ABLE, FIBWR, SIHULATE-E and CALTIP.

The NORGE-8 Cross-Section Representation Program Tabulation of cross section information for use by SIMULATE-E is a complex operation.

In the Supply System's methodology, CASMO-2 is first run for three void level depletion cases (01.,

401.,

and 701.

voids) and one con-trolled depletion at

01. void.

For the voided depletions, additional runs are made for instantaneous void branch

cases, Doppler

cases, controlled

cases, and cold cases.

For the controlled depletion, branch runs are made for uncontrolled hot and cold cases.

The resulting output is read by NORGE-8 which arranges the 'data into a

form recognizable by SIMULATE-E.

The SIMULATE-E formulation uses'he nine basic cross sec-'

tions shown in Table 3.1.

In the core simulation, these cross sections are made up of linear contributors to the t

overall cross section value and are termed partial cross sections.

Each partial cross section generated by NORGE-8 from the CASMO-2 data is the product of a tabulated func-tion of two variables and a polynomial function of a third variable.

NORGE-8 corrects a slight difference in k-infinity formula-tion between CASMO-2 and SIMULATE-E.

Under input control, the code can be made to adjust the fast group absorption cross section to make the two group nodal k-in'finitfes the same for both models.

This option is selected for deple-tion and core follow analyses but bypassed for the calcula-tion of kinetics constants and cross sections for system transient analyses.

Cross sections are formulated separately for hot and cold conditions to economize on computer memory requirements.

The accuracy of the NORGE-B cross section representations is assessed by comparing the results to the original CASNO cross section data.

The ABLE Albedo and Boundary Leakage Evaluation Program Reflective boundary conditions at the core periphery are calculated from the characteristics of the fuel bundles located along the edge of the core.

Specific albedo fac-tors are calculated for each peripheral bundle by the ABLE computer code ABLE uses core and fuel design information to calculate a

set of typical moderator conditions for the fueled and 0

non-fueled regions adjacent to the core periphery.

Truly typical conditions are used for the upper and lower core

faces, but each exterior fuel bundle vertical surface is provided with its own albedo evaluation.

The albedo calculation uses the lattice physics analysis for the moderator void fraction appropriate for the spe-cific albedo being evaluated.

Horizontal albedo boundary conditions are based on assembly-specific cross sections at an averaged void level.

Each calculated albedo is asso-ciated with a specific exposure distribution; because albe-does do not vary rapidly with exposure, average or typical 0

values are used in the Supply System methodology.

The SIMULATE-E Nodal Simulator Code The primary analytical tool used in core analysis is SIMULATE-E14.

Originally designed to allow the analyst to choose from a

number of calculational

schemes, SIMULATE-E has effectively evolved into an application of MCMDT.

The other nuclear analysis options are not used in the Supply System methodology.

Core analyses are performed in three dimensions using cubic

nodes, varying dimensionally only in thermal expan-sion effects.

In order to reduce computer

costs, these three-dimensional calculations normally use quarter core symmetry for depletion and power distribution calcula-

tions, relying on internal unfolding routines to generate full-core distributions when they are needed.

This approach generates a

small uncertainty in nodal quantities;

however, the benchmark results reported later in this chapter indicate that the magnitude of the errors introduced in this fash1on is small.

Input to SIMULATE-E includes cross

sections, fuel descrip-

tion, core description, core configuration, and calcula-tion control.

Cross sect1ons are provided by NORGE-8 on the bas1s of CASMO-2 analyses.

Fuel descr1ption informa-tion is taken from the appropriate design documents.

Core description information is taken from as-built design docu-ments and the results of supporting

analyses, such as those provided by ABLE and FIBWR discussed later.

Core configuration is determined on a case-specific basis and includes such variables as assembly loading

maps, control rod positions, total power, and power-flow state.

Calcula-tion control is also determined on a case-specific basis.

The Supply System's SIMULATE-E modeling was qualified through a

number of benchmark analyses.

Specific analyses selected for SIMULATE-E qualification include critical con-trol rod positions, incore instrument

response, and nodal power distribution.

The information against which these analyses were benchmarked was obtained from several sources.

These sources are core operating data covering the first two cycles of Quad Cities Unit 1, core operating data covering the first two cycles of Peach Bottom Unit 2, and operating data from four cycles of WNP-2.

Full SIMULATE-E depletion analyses were required to support these benchmark programs; the data allowed direct bench-marking of the SIMULATE-E analysis.

These benchmark analy-ses are summarized in Section 3.3, below.

3.1.4 The FIBHR Hydraulic Model SIMULATE-E contains a sophisticated flow-distribution algo-rithm that allocates core flow among the fuel assemblies in the core through equalization of assembly pressure drops.

This algorithm is an adaptation of the one con-tained in the FIBHR computer code

~

which uses the internally calculated bundle power distribution rather than a

zone distribution as in FIBHR.

With the exception of power distribution and inlet orifice information, the SIMULATE-E hydraulic input 'coincides with that specified for FIBWR.

Because of the complexity of the nodal power calculation, FIBHR is used in the Supply System's methodology to develop the hydraulic input for SIMULATE-E.

The flow dis-tribution in a SIMULATE-E calculation is. determined by the FIBWR model in SIMULATE-E.

FIBNR inputs are calculated from geometric design data.

A hydraulic calculation is performed with FIBNR using a

representative power distribution and results are compared with plant operating data.

Input formulations are adjusted until the FIBNR predictions and plant data agree; the SIMULATE-E input is then based on the adjusted FIBNR input.

3.1.5 The CALTIP Instrument

Response

Model The CALTIP computer program was developed by the Supply System to calculate TIP readings and compare them against measured Traversing Incore Probe (TIP) traces.

Calculated detector signal rates are based on fuel lattice type, expo-

.sure, void, control rod position and power.

'Oetector sig-nal rates are calculated by CASMO-2.

The contribution to the detector signal from a given assembly node is deter-mined by linear interpolation on detector signal rate using void and exposure values extracted from a SIMULATE-E restart file.

The detector signal rates at each node are converted to TIP traces by applying control rod worth and power dependencies.

The TIP signal is found by summing the rates from the four adjacent assembly nodes.

The sig-nals are normalized for each TIP string, normalized over the whole core, and reported.

3.2 HNP-2 BENCHMARKS Three different versions of SIMULATE have been used for core model-ing activities at HNP-2.

Some preliminary core modeling work was based on SIMULATE-1, with cross sections calculated by CPM.. The Supply System converted core modeling to SIMULATE-2 before startup of Cyc 1 e 1,

and the project changed again to SIMULATE-E for the third and subsequent cycles.

The CPM-based cross sections were used during the first operating

cycle, but the core was redepleted with CASMO-2 based cross sections when the lattice physics conversion was implemented.

The differences in the SIMULATE versions are largely peripheral, although SIMULATE-E includes the FIBNR hydraulic model-ing that was absent in the early versions.

Each operating cycle was modeled using the methods described in this report.

The results are summarized in this section; Figure 3.1 shows the performance of the SIMULATE critical eigenvalue as a function of core average exposure.

This figure contains data for both hot and cold critical predictions.

The trend toward increasing eigenvalue as core average exposure accumulates is con-sistent with other analysts'xperience with CASMO-2 and its inter-nal cross section library based on ENDF/B-III.

The consistent bias between hot and cold eigenvalues is also seen in other applications of CASMO-2.

The step increase in eigenvalue seen between Cycles 2

and 3

can be attributed mainly to different horizontal albedoes in the two cycles:

Cycles 1

and 2 used horizontal albedoes established for Cycle 1

which had natural uranium on the periphery.

Cycle 3

albedoes were determined with the use of the ABLE code modeling of the actual fuel residing on the core periphery during Cycle 3.

Com-parison of Cycle 3 results using the Cycle l albedoes to those using the ABLE code for Cycle 3

show a increase corresponding to that seen in Figure 3.1.

Critical rod positions were predicted with SIHULATE-E at a

number of cold statepoints during HNP-2 plant operation.

The results of these cold critical predictions are given in Table 3.2.

The cold critical eigenvalues exhibit a

general increase with increasing exposure.

For Cycle l

the mean k-effective is 0.999268 with a

standard deviat1on of 0.00189, while for Cycle 4

the mean k-effective is 1.00767 with a standard deviation of 0.00163.

Hithin a local range of exposures, cold k-effectives are predicted within a standard deviation of less than 2 mk.

Figures 3.2 through 3.33 show the results of HNP-2 TIP prediction benchmarks.

Each reported TIP dataset contains a

figure of the radial comparisons.,

showing how closely the calculated read1ngs agree with the measured readings on a core-normalized

basis, and a

composite figure showing core average axial readings and individual instrument string results for Instrument Locations 08-17, 56-17, 48-25, and 40-33.

The locations chosen for individual reporting were selected at random and include typical peripheral and internal core locat1ons.

For individual tip string axial comparisons, calcu-lated and measured results were normalized to one for each string.

Figures 3 '

through 3.9 show the results of the Supply System's core follow program for Cycles 1

and 2 of WNP-2.

While these TIP predic-tions were run with SIMULATE-2, they still demonstrate reasonable accuracy.

Cycle 3 results, based on the use of SIMULATE-E, are given in Figures 3.10 through 3.21.

Cycle 4 results are shown in Figures 3.22 through 3.33.

The analytical accuracy of these instru-ment predictions is summarized in Table 3.3.

The table includes all datasets at which TIP data are available.

The "RMS" column in Table 3.3 lists Drms (See Equation 8

in Section 2.2.2) calculated using all data points from all 43 TIP strings with calculated and measured datasets both normalized to one.

The range of RHS values shown is typical of that obtained by other licensees in compari-sons of SIMULATE calculations to neutron TIP data.

3.3 PEACH BOTTOM BENCHHARKS Transient and stability tests were run at Peach Bottom Unit 2 follow-ing the end of Cycle 2 operation EPRI published core design and operating data for the two cycles leading up to the tests.

This allows BWR licensees to determine the core conditions at the start of the tests.

An unanticipated benefit of this information was the issuance of data suitable for benchmarking reactor physics models.

The Peach Bottom modeling covers only hot conditions because the pub-lished data cover only power operation and test conditions.

The operating data consists of statepoint and configuration data for 24 separate data points during Cycle 1

and thirteen more during Cycle 2.

TIP readings were taken at many of these statepoints.

The published data indicates that all TIP datasets were taken after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of equi 1 ibrium operation.

The published power history data,

however, does not clearly support the equilibrium xenon

)udgment in all cases.

Table 3.4 lists all of the Peach Bottom statepoints.

Statepoints which could not be clearly fudged as equilibrium cond)-

tions are labeled as "xenon transient."

TIP calculations were not performed at these statepoints.

Quarter core geometry was adequate for the NNP-2 and Quad Cities benchmarks.

However, the Peach Bottom Unit 2

core was loaded with sufficient asymmetry that quarter-core averages were not thought to provide an adequate prediction of.localized effects.

For this seg-ment of the qualification work, the model was exercised in a full core mode.

The performance of the SIHULATE-E critical eigenvalue as a function of core average exposure during the first two operating cycles at Peach Bottom 2 is shown in Figure 3.34.

Some fluctuation in the'cal-culated eigenvalue is apparent at the end of the first cycle; this phenomenon can be attributed to significant levels of bypass coolant voiding following plugging of the bypass flow holes near the end of Cycle 1.

The introduction of bypass-enhanced fuel at the first refu-cling outage improved the bypass voiding conditions and the eigen-value restabilizes at that point.

Table 3.5 shows the rms TIP prediction error for each of the selected benchmark cases.

The TIP prediction error is the RHS-averaged error between predicted and measured instrument trace inte-grals for the whole core, calculated as described in Section 2.2.2, Equation (8).

The cycle averages shown in the table are the rms of the individual rms values.

The relative contributions of experimental uncertainty and calcula-tional uncertainty to the rms at each data point can be estimated by analyzing the measured and calculated results for assemblies which are symmetrically located about the SW-NE diagonal.

The rms aver-ages of these results show that the overall 11.3%

rms for Cycle 1

consists of approximately 4.91.

experimental rms and 10.27.'alcula-tional rms.

Similarly, the 8.41. overall rms for Cycle 2 consists of 5.81.

experimental rms and 6.01. calculational rms.'he overall rms for Cycle 1

has a

greater contribution from calculational

error, probably reflecting the modeling difficulties associated with the plugging of the bypass flow holes.

The results for Cycle 2

show a

lower overall rms containing approximately equal contributions from experimental and calculational uncertainties.

Figure 3.35 shows core average TIP prediction error as a function of core average exposure.

Figures 3.36 through 3,47 provide core aver-age and selected individual TIP instrument benchmark comparisons for typical TIP datasets collected during Cycle 1.

Figures 3.48 through 3.55 show.core average and selected individual TIP instrument bench-mark comparisons for typical statepoints during Cycle 2.

The instru-ment response figures are consistent with those shown for WNP-2.

These plots indicate a

reasonable prediction of axial and 'radial power shape trends over the two operating cycles which compose the benchmark.

3. 4 QUAO CITIES BENCHMARKS The Quad Cities Unit 1

core desi.gn and operating data includes descriptions of the initial core and reload fuel, operating system parameters for hot and cold state points, and TIP data for the first two cycles.

Also, in addition to local power measurements, the Quad Cities gamma scan measurements cover nodal Lanthanum-140 distribu-tions accross the core at the end of Cycle 2.

Quad Cities Unit 1

was modeled in quarter core geometry.

As in the Peach Bottom analysis, there were some asymmetries i,n core loading and control rod patterns, but the severity was not as great.

The quarter core analysis did result in some difficulty in predicting instrument responses in the quadrants of the core which were not modeled explicitly.

For the second

cycle, six reported instrument strings were not predicted because fuel assemblies adjacent to those instruments were loaded asymmetrically from their counterparts in the analyzed quadrant.

The Quad Cities Unit 1

core was depleted from beginning of Cycle 1

to end of Cycle 2 in discrete steps corresponding to the distribu-tion of formal datasets in the 'literature.

The specific statepoints selected for the Quad Cities analysis are shown in Table 3.6.

The general performance of the core model is shown in Figure 3.56, which provides the SIMULATE-E critical ei genvalue as a function of core average exposure.

The Quad Cities work shows results similar to the Peach Bottom analysis'he cold cross section model was used at a

number of statepoints throughout the analysis to predict the cold criticals published in the data.

The cold critical predictions are summarized in Table 3.7.

The two results columns in the table list the actual eigenvalue calculated by SIMULATE-E and the adjusted value obtained by applying a correction for reactor period.

The performance of the Supply System's model in predicting the pub-lished TIP data is shown in Table 3.8 and Figure 3.57.

The results are again similar to the Peach Bottom resul.ts, with the Cycle 2 pre-dictions generally more accurate and stable than those of Cycle l.

Figures 3.58 through 3.67 provide typical Cycle 1

TIP predictions, while Figures 3.68 through 3.75 show typical results for Cycle 2.

These results are shown in the manner used to present the HNP-2 and Peach Bottom results, but with the individual instrument strings selected differently.

In the Quad Cities benchmarks, individual instrument predictions are reported for-Instrument Locations 08-17, 08-25, 56-25, and 48-33 during Cycle 1

and for Instrument Locations 16-17, 32-25, 40-33, and 32-41 during Cycle 2.

The results shown in Table 3.8 yield a Cycle 1 overall rms of 10.7'I and a Cycle 2 overall rms of 10.0%%d.

Estimates of the contributions of measurement uncertainty and calculational uncertainty to these overall rms's were obtained by analysis of asymmetry as described in Section 3.3.

The results show that for Cycle 1,

the experimental rms is 7.61.

and the calculational rms is 7.5'/, while for Cycle 2 the experimental rms is 6.31.

and the calculational rms is 7.71..

The overall uncertainty factors contain approximately equal contribu-tions from experimental and calculational uncertainties.

The nodal power benchmarks also include comparisons of normalized calculated and measured Ba-140 distributions.

The Ba-140 concentra-

tions, which are the determining mechanism for the La-140 activity during the gamma scan

period, were benchmarked in both radial (x-y) and axial dimensions.

The radial gamma scan benchmarks are based on calculated and mea-sured activities normalized to one over the entire set of 89 scanned assemblies.

Normalization of the measured data required some care since assemblies could have either 12 point scans or 24 point scans, and some scan points were displaced 1-2 inches from nodal centers.

The measured data for each assembly was fit by a

natural cubic

spline, and the assembly average activity was defined in terms of the analytical integral of the spline.

44

Of the 89 fuel assemblies scanned for La-140 activity, 77 were physi-cally located in the west-northwest octant of the core or near the core centerline.

Figure 3.76 shows the assembly averaged Ba-140 con-centration comparisons for this core region, which contains the bulk of the data.

Drms for this comparison is 2.37'A.

Figures 3.77 through 3.79 provide typical radial power benchmarks at selected axial levels in the core.

Figure 3.77 shows the results.for Plane 7,

which has the largest RMS difference, Drms 4.1'I.

Figure 3.78 shows Plane 15, which is more typical with Drms-3.1'l.

Figure 3.79 shows Plane 18, which has the smallest

Orms, 1.6X.

For these radial comparisons, the calculated and measured results were normalized to one on each axial plane.

The normalizations included only those nodes at which measured data exits.

The uncertainties in the axial power shape were evaluated separately, as described below.

The rad-ial benchmarks for all 24 axial planes are summarized in Table 3.9.

The axial power gamma scan benchmarks are summarized in Table 3.10.

This table contains the rms difference between measured and calcu-lated intensity values divided by the axial average measured inten-sity for the assembly.

Measured and calculated results are normalized to one over the entire set of scanned assemblies.

Calcu-lated activities at nodes with off-center scan points were obtained by spline interpolation of the calculated nodal center activities.

Calculation of rms differences excluded data from the top and bottom six-inch nodes.

Figure 3.80 shows the core average axial distribu-tion benchmark.

Figures 3.81 through 3.84 provide axial power bench-marks for selected assemblies.

The calculated and measured axial peak-to-average activiti'es for each assembly were also compared and the results are shown in Table 3.11.

This table shows the percent differences, ei, calcu-lated as in Section 2.2.2, Equation (1).

The overall Erms for all 89 assemblies, calculated by Equation (7) of the same

section, is 1.53'/,

and the average error is 0.591..

TABLE 3.1 HNP-2 CROSS SECTION FORMULATION ctrl Exposure Instantaneous Void Eal Erl vEgl KEgl Exposure Control Instantaneous Void Void History Fuel Temperature Control History Exposure Control Instantaneous Void Void History Exposure Instantaneous Void Void History Exposure Instantaneous Void Void History Exposure Instantaneous Void Ea2 vEf:2 xEf:2 Exposure Control Instantaneous Void Void History Control History Xenon Number Density Samarium Number Density Exposure Control Instantaneous Void Void History Control History Fuel Temperature (Cold model only)

Exposure Control Instantaneous Void Void History Control History TABLE 3.2

SUMMARY

OF NNP-2 COLD CRITICAL PREDICTIONS

'I 0

0 0

57 121 192 268 440 661 1037 2337 2502 3253 3799 5929 7065 7065 7167 "10209 11661 9650 9650 12666 14022 14383 11052 12368 14173 Cycle 1

Cycle 2

Cycle 3

Cycle 4

1.0013 1.0008 0.9999 1.0008 1.0014 1.0019 1.0004 0.9998 0.9989 0.9989 0.9971 0.9975 0.9973 0.9969 0.9962 0.9957 0.9965 0.9961 0.9967 0.9997 1.0007 1.0029 1.0058 1.0061 1,0077 1.0059, 1.0091 1.0081 TABLE 3.3

SUMMARY

OF WNP-2 TIP DATASETS Core Average EEU5~

TIP Prediction

~II 44 57 64 93 Cycle 1

1372 2205 2590 4579 5.09 6.45 9.57 15.40 4

10 17 26 34 40 Cycle 2

7342 8248 9297 10360 11339 12102 6.59 6.40 6.15 6.09 7.55 8.49 1

7 8

15 20 27 Cycle 3

9903 10836 11034 12245 12903 13974 7.89 7.97 8.38 9.75 8.47 5.43

'1ll 17 23 30 37 Cycle 4

11245 12903 13563 14428 15473 16531 5.73 11.05 7.87 6.90 5.92 6.54 TABLE 3.4 PEACH BOTTOM 2 CYCLE 1&2 STATEPOINTS Dataset Hua@e Core Average Exposure

~D~

~~/

Cycle 1 Data Data Exalt~

1 2

3 4

5 6

7 8

9 10ll 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 230 439 648 741 1010 1585 2080 2555 2920 3724 4525 4697 5262 5640 6106 6470 7000 7300 7712 8100 84.30 8766 9520 10100 8042 8706 9500 9730 10050 10730 11030 11260 11420 11570 11910 12190 12530 253 484 714 817 1113 1747 2293 2816 3219 4105 4988 5178 5800 6217 6731 7132 7716 8047 8501 8929

'9292 9663 10494 11133 Cycle 2 Data 8865 9597 10472 10726 11078 11830 12159 12412 12588 12754 13129 13437 13812 Xenon transient Xenon transient No TIP data Xenon transient Modeled No TIP data No TIP data Hodeled Modeled Xenon transient Xenon transient Modeled Modeled Modeled Modeled Modeled No TIP data No TIP data No TIP data Modeled No TIP data Modeled Xenon transient Modeled Modeled Xenon transient Xenon transient Model ed Modeled Modeled Modeled Xenon transient Xenon transient Modeled Modeled Xenon transient Modeled Dataset gi~e~

TABLE 3.5 PEACH BOTTOH 2 CYCLE 1&2 BENCHMARKS TIP Prediction Error Cycle 1 Data 5

8 9

12 13 14 15 16 20 22 24 6.9 12.3 15.6.

13.1 9.0 7.9 8.5 9.8 19.2 7.8 7.3 Cycle 1 Overall Drys 11.31.

25 28 29 30 31 34 35 37 Cycle 2 Data 9.3 8.5 8.3 8.6 8.6 8.3 7.8 7.4 Cycle 2 Overall Drys 8.41.

Table 3.6 squad Cities 1 Statepoints Dataset ggg~r Core Average Exposure

~D~

~D~

Cycle 1 Data Data 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16

~ 17 18 19 20 21 22 23 24 25 26 27 28 29 247 646 800 1334 2031 2894 3480 3696 4297 4809 5471 5949 6175 6710 6948 7239 6625 6833 7225 7641 7973 8293 9229 10195 10827 11699 11973 12348 12466 272 712 882 1470 2239 3190 3836 4074 4737 5301 6031 6558 6807 7397 7659 7980 7303 7532 7964 8423 8789 9141 10173 11238 11935 12896 13198 13611 13741 Cycle 2 Data Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Modeled Table 3.7 Quad Cities Cycle 152 Benchmark Summary of Cold Critical Predictions Core Average Exposure W

T SIMULATE-E 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17 18 19

~

20 21 22 23 24 25 26 26 28 29 30 31 33 34 36 37 C5 C6 C7 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 2866 2866 3748 3748 3748 3748 3748 3748 3748 3748 4938 6912 6912 8276 9177 10658 Cycle 1 Data 1.0000 1.0005 0.9988 0.9998 0.9999 0.9987 0.9997 0.9997 0.9999 1.0027 0.9997 1.0001 1.0003 1.0027 1.0027 1.0027 1.0027 1.0028 1.0030 0.9993 1.0008 0.9989 0.9960 0.9961 1.0003 1.0001 0.9990 0.9977 0.9960 0.9958 0.9958 0.9961 0.9959 0.9992 0.9968 Cycle 2 Data 0.9981 0.9991 0.9988 0.9996 0.9995 0.9980 0.9993 0.9994 0.9976 0.9982 0.9988 0.9991 1.0014 0.9988 0.9995 0.9994 1.0025 1.0024 1.0013 1.0015 1.0014 1.0016 0.9989 1.0004 0.9983 0.9957 0.9958 0.9991 0.9990 1.0001 0.9966 0.9958 0.9953 0.9953 0.9957 0.9946 0.9985 0.9966 0.9974 0.9979 0.9982

• Data sheet 26 gives two sets of data.

Tabl e 3.8 Dataset Humtzz 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 Quad Cities 1 Cycle 152 Benchmark Summary of TIP Predictions TIP Prediction Error

~orms Cycle 1 Data 9.6 8.9 8.6 10.2 9.3 10.3 11.1 11.8 10.0 9.4 13.0 9.9 9.6 11.5 10.9 14.6 Cyc 1 e 1 Overall Drms

10. 7'L 17 18 19 20 21 22 23 24 25 26 2728-29 Cycle 2 Data

10. 4 10,3 9.6 10.6 9.4 9.8 9.9 8.3 10.1 9.0 8.4 11.0 12.1 Cycle 2 Overall Drms

10. 0'/

Axial

$~1 24 23

-22 21 20 19 18 17 16

~ 15'4 13 12ll 10 9

8 7

6 5

4 3

2 1

Table 3.9 Quad Cities Gamma Scan Benchmark Summary of Radial Power Benchmarks Dr s

4. 6/.

3. 0'/

2. 5'/

2. 5'/

1. 9/.

2. 5'/

1. 6'X

2. 5'/

2. 4/

3. 1'/

2.3/

3.0X

3. 0%

3. 3'4

3. 7'4 3.8/

3. 9/.

4.1/

3.4/

3.9/

2:9/.

3. 2'/

2. 1'/

5.1/

Table 3.10 Quad Cities Gamma Scan Benchmark Summary of Axial Power Benchmarks

~Drms

~Drms CX0546 CX0191 CX0494 GEH008 CX0412 CX0174 CX0100 CX0553 GEH029 CX0662 CX0440 CX0378 CX0682 CX0351 CX0396 CX0316 GEB149 CX0399 CX0231 CX0297 CX0523 GEH022 GEH002 GEB132 CX0281 CX0643 CX0327 CX0306 CX0287 CX0310 CX0214 CX0520 CX0420 CX0394 CX0052 CX0717 CX0453 CX0482 GEB159 GEB158 CX0318 CX0096 CX0384 CX0124 CX0622 9.09 6.28 6.88 4.89 4.97

6. 95 7.61 11.96 4.30 8.33 7.55 5.07 7.91 6.53 4.90 6.34 4.19 7.78 4.63 3.79 5.84 6.91 4.16 5.02 4.27 10.66 4.99 4.13 3.76 3.65 2.88

5. 10 5.88 4.26 4.35 3.79 3.48 4.40 3.64 3.99 4.02 3,88 6.37 6.54 9.81 CX0719 GEB162 CX0286 CX0398 CX0490 CX0617 CX0024 CX0332 GEB123 CX0150 CX0015 CX0186 CX0611 GEH023 CX0093 CX0723 CX0631 CX0397 CX0161 CX0414 CX0198 CX0393 CX0672 CX0585 GEB105 CX0044 CX0362 CX0660 CX0498 CX0575 CX0683 CX0137 CX0106 CX0057 CX0162 CX0225 CX0165 GEB161 CX0588 GEB160 CX0401 CX0445 CX0359 CX0711 7.15 9.02 7.44 9.24 10,96 7.69 6.61 6.92 5.09 7.18 6.13 4.77 9.06 6.32 5.92 4.52 6.78 6.00 4.07 3.92 5.35 4.78 4.46

'.90 4.91 5.00 4.55 4.16 3.77 2.92 10.63 4.42 4.49 4.66 3.81 3.70 3.50 3.86 3.98 3.05 5.76 7.15 8.32 7.87 Table 3.11 Assembly Quad Cities Gamma Scan Benchmark Axial Peak-to-Average Comparisons Axial Peak-to-Average Hazmzed CX0546 CX0719 CX0191 GEB162 CX0494 CX0286 GEH008 CX0398 CX0412 CX0490 CX0174 CX0617 CX0100 CX0024 CX0553

. CX0332 GEH029 GEB123 CX0662 CX0150 CX0440 CX0015 CX0378 CX0186 CX0682 CX0611 CX0351 GEH023 CX0396 CX0093 CX0316 CX0723 GEB149 CX0631 CX0399 CX0397 CX0231 CX0161 CX0297 CX0414 CX0523 CX0198 GEH022 CX0393

1. 372 1.331 1.302 1.353 1.341 1.314 1.212 1.203 1.203 1.344 1.324 1.287 1.222 1.209 1.331 1.217 1,183 1,183

1. 388 1.284 1.229 1.191 1.210 1.225 1.417 1.359 1.251 1.198 1.197 1.196 1.215 1.213 1.199 1,308 1.215 1.198 1.206 1.222 1.213 1,198 1.343 1.295 1.201 1.213 1.354 1.331 1.293 1.355 1.352 1.324 1.212 1.194 1.185 1.354 1,337 1.307

1. 221 1.188 1.354 1.218 1.168 1.167 1.378 1.308 1.257 1.189 1.212

1. 205

1. 416

1. 390

1. 270

1. 202 1.206 1.218

1. 216

1. 202 1.187

1. 323 1.244 1.210 1.218 1.223 1.198 1.148 1.345 1

~ 284 1.196 1.203 1.28 0.00 0.68

-0.18

-0.81

-0.79

-0.01 0.76 1.53

-0.70

-0.92

-1. 51 0.08 1.77

-1. 70

-0.03 1.31 1.35 0.72

.1.86

-2.26 0.18

-0.16 1.70 0.11

-2.25

-1.45

-0.26

-0.73

-1.84

-0.09 0.95 0.97

-1.10

-2.36

-1.02

-1.01

-0.13 1.27 4.36

-0.13 0.80 0.42 0.82 Table 3.11 (Continued)

Assembly n ifi i

n Quad Cities Gamma Scan Benchmark Axial Peak-to-Average Comparisons Axial Peak-to-Average

~11 hI GEH002 CX0672

'GEB132 CX0585 CX0281 GEB105 CX0643 CX0044 CX0327 CX0362 CX0306 CX0660 CX0287 CX0498 CX0310 CX0575 CX0214 CX0683 CX0520 CX0137 CX0420

'X0106 CX0394 CX0057 CX0052 CX0162 CX0717 CX0225 CX0453 CX0165 CX0482 GEB161 GEB159 CX0588 GEB158 GEB160 CX0318 CX0401 CX0096 CX0445 CX0384 CX0359 CX0124

'X0711 CX0622 1.179 1.199 1.171 1.192 1.190 1.172 1.364 1.257

1. 241

1. 220

1. 205 1.186 1.188 1.203

1. 210 1.199 1.210 1,361 1.337 1.289 1.261

'1. 255 1.248 1.231 1.203 1.184 1.176 1.185 1.198 1.215

1. 206 1.148 1.148 1.206 1.148 1.148 1.198 1,198 1,251 1.210 1.203 1.203 1.224 1.314 1.364 1.168 1.198 1.160 1.211 1.184 1.167 1.332 1.242 1.227 1.213 1.200 1.186 1.191 1.187 1.185 1.196 1.195

1. 326

1. 308 1.275 1.239 1.233 1.224 1.222 1,196 1.178 1.155 1.185

1. 200 1.191 1.178

1. 137 1.141 1.187 1.133 1.138 1,156 1.216 1.201 1.201 1.193 1.167 1.215 1.288 1.309 0.95 0.12 0.99

-1.57 0.52 0.42 2.38 1.24 1.13 0.58 0.46 0.03

-0.22 1.30 2.13 0.23 1.21 2.62 2.18 1,08 1.76 1.74 1.96 0.81 0.56 0.55 1.79

-0.03

-0.19 2.02 2.46 1.01 0.63 1.62 1.34 0.87 3.66

-1.50 4.12 0.77 0.81 3.04 0.74 2.06 4.23 Figure 3.'I MfNP 2 Criticai Eigenvalue Calculated by SIMULATEE 1.02 1.01 1.00

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2,500 5,000 7,500 10,000 Core Average Exposure (MWd/MT) 12,500 15,000

WNP-2 Cycle Figure 3.2 1 TIP Predictions, 1372 MWd/HT 16-57 0.599 0.579

-3.252 24-57 0.852 0.856 0.421 32-57 0.888 0.882

-0.670 40-57 0.777 0.776

-0.094

+

08-41 0.986 0.998

1. 201 16-41 24-41 1.083 1.073 1.101 1.076 1.078 0.301 08-49 16-49 24-49 0.700 1.021 1.057 0.652 1.030 1.100

-6..759 0.938 4.084 32-49 1.115 1.113

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~ 114

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-1. 813 40-33 1.274 1.252

-1.759 48-33 1.113 1.136 2.098 56-33 0.815 0.790

-3.019 08-25 0.986 1.015 3.007 16-25 1.177 1.152

-2.099 24-25 1.082

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48-25 1.133 1.126

-0.688 56-25 0.815 0.788

-3.303 08-17

- 0.937 0.918

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-0.965 40-17 1.062 1.072 0.944 48-17 1.098 1.090

-0.732 56-17 0.619 0.577

-6.789 16-09 0.933 0.914

-2.026 24-09 1.098 1.123 2.197 32-09 1.135 1.147 1.032 40-09 1.048 1.059 1.033 48-09 0.714 0.656

-8.227 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.3 WNP -2 Cycle 1 TIP Predictions, 1372 MWdiMT Core Average Instrument Response tassel

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Fi gure 3.4 NNP-2 Cycle 1 TIP Predictions, 4579 HHd/MT 16-57 0.629 0.581

-7.492 08-49 16-49 0.721 1.078 0.668 1.075

-7.358

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-7.641 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.5 liYNP-2 Cycle 1 TIP Predictions, 4579 MWdlMT Core Average lnatrument Reaptpnae 4411444}

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Figure 3..6 WNP-2 Cycle 2 TIP Predictions, 9297 MWd/MT 08-49 0.788 0.723

-8.293 08-41 1.060 1.036

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+

24-57 0.962 0.876

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1. 380 24-33 1.038 1.077 3.774 24-25 0.998 1.075

7. 685.

32-57 0.932 0.888

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+

40-41 1.021 1.074 5.163 40-33 1.063 1.087 2.205 40-25 1.048 1.105 5.403 48-49 1.026 1.006

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1. 06'1

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• -6.786 08-17 0.970 0.940

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3. 901 24-.09

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2. 614 40-09 1.023 1.026 0.338 48-17 1.051 1.089 3.583 48-09 0.813 0.721

-11.352 56-17 0.741 0.658

-11.174 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.7 WNP 2 Cycle 2 TIP Predictions; 9297 MWd/MT Core Average Instrument Response 4'.I

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Figure 3.8 HNP-2 Cycle 2 TIP Predictions, 12102 MHd/MT 08-49 0.738 0.718

-2.615 16-57 0.697 0.665

-4.593 16-49 1.088 1.104 1.488 24-57 0,898 0.830

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-8.319 40-49 1.113 1.101

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-1

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-1.350 40-41 48-41 0.950 1.138 1.020 1.103 7.316

-3.072 56-41 0.871 0.806

-7.448 56-33 0.890 0.831

-6.719 08-25 1.006 1.035 2.919 16-25 1.001

" 1,032 3.072 24-25 1.098 1.172 6.723 32-25 1.139 1. 2'1 9 7.091 40-25 1.021 1.089 6.646 48-25 1.085 1.071

-1.303 56-25 0.897 0.830

-7.422 08-17 0.995 0.982

-1. 347 16-17 1.038 1.069 3.055 16-09 1.052 0.981

-6.705 24-17 0.988 1.031 4.334 24-09 1.006 1.034 2.854 32-17

0. 997 1.032 3.436 32-09 1.022 1.003

-1.863 40-17 48-17 56-17 1.055

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-1.509

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-5.924 -11,452 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.9 YVNP 2 Cycle 2 TIP Predictions, 12102 MMfdlMT Core Averege Instrument Response

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-2.887 08-41 1.122 1.116

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-3.783 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

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-1.489 XX-XX X.XXX X.XXX X.XXX

+

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String Identification Measured Average Instrument

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-4.938 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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Calculated Average Instrument

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-3.858 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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-4.581 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

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-6.343 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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-0.458 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

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-4.540 24-41 1.189 1.133

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1. 131

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-0.190 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.27 WNP 2 Cycle 4 TIP Predictions, 13563 MWdlMT Core Average Instrument Response

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1. 107

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-0.007 56-17 0.654 0.640

-2.270 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

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. Figure 3.29 WNP-2 Cycle 4 TIP Predictions, 14428 MWdlMT Core Average Insfrumenf Response 2.00 8 1.50

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+

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1. 117

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1. 358

4. 194 XX-XX X.XXX X.XXX X.XXX String Identification Heasured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.31 WNP -2 Cycle 4 TIP Predictions, i5473 MLVdiMT Core Average Instrument Response

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-0.912 40-49 1.126 1.065

-5.437 40-41 1;063 0.994

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-7.500 40-25 1.088 1.010

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'.788 0.650 56-17 0.667 0.693 3.863 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error Figure 3.39 Peach Bottom 2 Cycle 1 TIP Predictions 2616 MWd/MT Cote Atterage Inatnlment Response

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+

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1. 010

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1. 023 1.037 1.315 24-33 1.060 1.014

-4.343 32-33 1.097 1.016

-7.417 40-33 1.083 1.013

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-4.074 56-33 0.795 0.887 11.541 08-25 1.144 1.139

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1. 047

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Response

Calculated Average Instrument

Response

Percent error Figure 3.4 I Peach Bottom 2 Cycle 1 TIP Predictions 5178 MWdlMT Core Aeerage Instrument Response (Zht

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1. 100 1.589 08-33 1.078 1.102 2.232 16-41 1.090 1.058

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56-33 0.797 0.876 9.808 08-25 1.139 1.118

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-5.815 40-09 1.094 1.097 0.256 48-17 1.062 1.046

-1.495 48-09 0.791 0.794 0.396 56-17 0.689 0.694 0.815 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error

-100-

Figure 3A3 Peach Bottom 2 Cycle 1 TIP Predictions 5800 MNdlMT Cole Average Instrument Response IVVV

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Response

Calculated Average Instrument

Response

Percent error

-102-

Figure 3.45 Peach Bottom 2 Cycle 1 TIP Predictions 6731 MNfd/MT Core Average Instrument Response CS C

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Peach Bottom 2 Cycle Figure 3.46 1 TIP Predictions, 11133 MWd/MT 16-57 0.661 0.588

-11.072 24-57 32-57 0.808 0.826 0.796 0.840

-1.413 1.739 40-57 0.855 0.773

-9.620 08-49 0.745 0.679

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1. 268 16-33 1.029 1.164 13.189 16-25 1.101 1.156 5.000 16-17 1.108 1.090

-1. 597 24-41 32-41 1.036 1.067 1.147 1,114 10.707 4.429 24-33 1.044 1.126 7.806 32-33 1.120 1.130 0.902 24-25 32-25 1.168 1.115 1.145 1.119

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-1.437 40-25 1.179 1.158

-1.763 40-17 1.106 1.124

1. 704 48-41 1.053 1.079 2.556 48-33 1.118 1.149 2.770 48-25 1,119 1.117

-0. 191 48-17 1.062 1.021

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-8.589 16-09 0.861 0.882 2.362 24-09 1.012 1.028 1.565 32-09 1.051 1.053 0.153 40-09 0.989 0.983

-0,588 48-09 0.766 0.679

-11.310 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

Response

Percent error

-104-

Figure 3.47 Peach Bottom 2 Cycle 1 TIP Predictions 17133 MWd/MT Core Average Instrument Response

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-6.329 24-57 0.934 0.966

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-15.274 56-17 0.771 0.728

-5.658 XX-XX X.XXX X.XXX X,XXX String Identification Measured Average Instrument

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Calculated Average Instrument

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-106-

Figure 3.49 Peach Bottom 2 Cycle 2 TIP Predictions l0726 MWd/MT Core A)>grege Instrument Response a

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-6.074 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

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-108-

Figure 3.51 Peach Bottom 2 Cycle 2 TIP Predictions i107S Myyd/MT Core A)serage Instr)))nant Response e

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-4.537 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

Response

Calculated Average Instrument

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-110-

Figure 3.53 Peach Boffom 2 Cycle 2 TIP Predictions l2754 MWd!MT Core Average Instrument Response jj,C.'441.'jj.l.

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Figure 3. 54 Peach Bottom 2 Cycle 2 TIP Predictions, 13812 MWd/HT 16-57 24-57 32-57 40-57 0.741 0.905 0.884 0.985 0.690 0.931 0.934 0.870

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+

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-0.228 16-41 24-41 32-41 1.088 1.042 1.020 1.099 1.036 1.024 1

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-14.330 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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-112-

Figure 3.55 Peach Bottom 2 Cycle 2 TIP Predictions 13S12 MWdlMT Core Atterage Inttrument Retponte F

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-4.843 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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-116-

Figure 3.59 Quad Cities 7 Cycle 1 TIP Predictions 772 MWd(MT Core Atrerege Instrument ResPonse

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X.XXX X.XXX String Identification Measured Average Instrument

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Figure 3. 62 Quad Cities 1 Cycle 1 TIP Predictions, 4737 MHd/MT 08-49 0.600 0.559

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-9.196 XX-XX X.XXX X.XXX X.XXX String Identif1 cation Measured Average Instrument

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-120-

Figure 3.63 Quad Cities 1 Cycle 1 TIP Predictions 4737 MWdlMT Core A))crepe Instrttmen! Response c

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Figure 3.64 Quad Cities 1

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-7.194 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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Figure 3.65 Quad Cities 1 Cycle 1 TIP Predictions 6807 MyVd/MT Co20 Awrnge In2trttment R02Ponne

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-9.315 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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-124-

Figure 3.67 Quad Cities i Cycle 1 TIP Predictions 806S MYYd/MT Core Atrerape Instrument Reeponee

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-7.323 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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Figure 3.70 Quad Cities 1 Cycle 2 TIP Predictions, 9141 MHd/MT 24-57 0.926 0.854

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-8.846 XX-XX X.XXX X.XXX X.XXX String Identification Measured Average Instrument

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Figure 3.71 Quad Cities 1 Cycle 2 TIP Predictions 9141 MMfd/MT Core Average Instrument Response c 1.50 IX cI 2 1.00

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-9.166 56-25 XX-XX X.XXX X.XXX X.XXX String Ident if1 cat i on Measured Average Instrument

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-130-

Figure 3.73 Quad Cities 1 Cycle 2 T/P Predictions 13198 MWdlMT Cotej Averege Instrument Response

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-6.014 XX-XX X,XXX X.XXX X.XXX String Identification Measured Average Instrument

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Calculated Average Instrument

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Figure 3.75 Quad Cities 1 Cycle 2 TIP Predictions 13741 MMfd/MT Core Average Inetrument Reeponee 20c 4.00 a

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Figur e 3. 76 Quad Cities Radial Power Benchmark Assembly Averaged Activity Levels 0.624 0.622 R.M 0 822

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Figur e 3. 77 Quad Citiea Radi.al Power Benchmark Radial La140 Oiatribution at Axial Plane 7

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Figur e 3. 78 Quad Cities Radial Power Benchmark Radial La14.0 Distribution at Axial Plane 15 0.658 0.681 3.42 0.876 1.071

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!IM CALC sQIF Norma]ization:

Sus of 77 bundles at this elevation set to 100.00.

1.705 1,675 1.635

].635

-4.13

-2.40 136

Figur e 3. 79 Quad Cities Radial Power Benchmark Radial La14.0 Distr ibution at Axial Plane 18 3.218 4.291 3.347 4.224 3.99

-1.57 3.491 4.898 3.683 4.983 5.48 1.74 I I I I I I I I I I 111 111 I I I I I

I I I I I I I I 1111111111111111 I 111 111 I

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I I I I I I I I I I I I I I I I I

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Sm of 16 bmlfes bt 'this efevetiorl set to 100.00.

8.023 7.961 7.933 7.933

-1.13 4.35 137

Figure 3.SO Core Average AxialLa -740 Distribution Quad Cities Gamma $can Benchmark I

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12 24 36 48 60 72 84 96 Distance Above Bottom of Active Fuel (Inch) 108 120 132 144

Figure 3.81 AxialLa -740 Distribution forAssembly CX0553 Quad Cities Gamma Scan Benchmark 1.00 0.75 EOCI C

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Figure 3.82 Axiai La -140 Distribution forAssembly GEH023

'uad Cities Gamma Scan Benchmark 2.00

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Figure 3.83 Axiai La -140 Distribution forAssembly CX0274 Quad Cities Gamma Scan Benchmark 2.00 1.50 COcIc

~~ 1.00 N

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Figure 3.84 AxialLa -140 Distribution forAssembly GEB161 Quad Cities Gamma Scan Benchmark 2.00 1.50 I 1 o N

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12 24 36 48 60 72 84 96 Distance Above Bottom of Active Fuel (Inch) 108 120 132 144

4.0

SUMMARY

AND CONCLUSIONS The benchmark comparisons presented in this topical report quantify the accuracy of the Supply System core physics model in steady state applica-tions and demonstrate the qualifications of the Supply System engineering staff to perform steady state core physics calculations in support of the WNP-2 nuclear plant.

An overall summary of the benchmarks presented in this report follows.

The lattice physics benchmarks include uniform lattice criticals and 4

assembly local power distributions.

Fifteen uniform lattice critical s, consisting of eight TRX U02 criticals and seven ESADA mixed oxide criti-

cals, were analyzed.

The eight TRX U02 criticals yielded a

mean k-effective of 0.99616 with a

standard deviation of 0.00184.

The seven ESADA mixed oxide criticals yielded a

mean k-effective of 1.00872 with a standard deviation of 0.00782.

The local.power distribution benchmark was based on gamma scans of individual fuel pins at the end of Cycle 2 at Quad Cities 1.

Comparisons of local powers of pins from five scanned assemblies yielded an 'overall standard deviation of 3.20K.

The published measurement uncertainty is 1.7X, indicating that calculational and meas-urement uncertainties are comparable.

Comparisons of maximum local peak-ing factors yielded an overall standard deviation of 2.14/

which is even closer to the measurement uncertainty of 1.71.

The core si.mulation benchmarks are based on data from WNP-2, Peach Bottom 2,

and Quad Cities 1.

They include hot and cold criticals, TIP compari-

sons, and nodal gamma scan comparisons.

-143-

4 The overall performance of the calculated k-effectives is shown in Figure 3.1 (WNP-2), Figure 3.34 (PB-2),

and Figure 3.56 (QC-1).

As observed by other licensees, the CASHO-2/SINULATE-E analysis produces k-effectives which show a general increase with increasing

exposure, and which exhibit a small bias between hot and cold criticals, The overall performance of the TIP comparisons is shown in Table 3.3 and Figures 3.2-3.33 (WNP-2),

Table 3.5 and Figures 3,35-3.55

.(PB-2),

and Table 3.8 and Figures 3.57-3.75 (QC-1).

The overall TIP rms for WNP-2 (4 cycles) is 8.081..

The corresponding results for PB-2 and QC-1 (two cycles) are 10.181.

and 10.65't, respectively, Analysis of asymmetries suggests that these uncer-tainties contain approximately equal contributions from experimental and calculational uncertainties.

An additional benchmark of the nodal power distribution is provided by the nodal gamma scan data from Quad Cities l.

The comparisons of calculated and measured activities are shown in Figures'.76-3.84 and in Table 3.10.

Comparisons of calculated and mea-sured peak-to-average activities for'ach assembly are shown in Table 3.11, which shows an average difference of 0.591. with an rms of 1.531..

In

summary, the results presented here demonstrate the ability of the Supply System staff to set up and run detailed steady state core physics calculations in support of'he WNP-2 plant.

The validity of the Supply System methodology and accuracy of the numerical results is demonstrated by extensive comparisons to measured data from WNP-2, Peach Bottom 2, and Quad Cities 1.

-144-

T Supply System management has a strong commitment to maintain and improve in-house core physics capabilities.

The initial development of these capabilities and their enhancement is part of the overall Supply System's objective of self sufficient design capability with respect to the ownership of WNP-2.

Ongoing core follow analysis and TIP comparisons per-formed in support of WNP-2 plant operations provide continuing opportuni-ties to benchmark Supply System core physics methods.

-145-

5.0 REFERENCES

2.

3.

4.

5.

6.

7.

8.

9.

10.

12.

"Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions,"

Generic Letter 83-11, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, dated February 8,

1983.

"Qualification of Steady State Core Physics Methods for BWR Design and Analysis," PL-NF-87-001, Pennsylvania Power

& Light Company (1987).

"Methods for the Analysi s of Boiling Water

Reactors, Steady State Core Physics,"

YAEC-1238, Yankee Atomic Electric Company (1981).

H.

Edenius and A. Ahlin, "MICBURN., Microscopic Burnup in Gadolinia Fuel Pins,"

EPRI CCM-3', Part II, Chapter 7, Electric Power Research Institute (1975).

H.

Edenius, A. Ahlin, and H.

Haggblom, "CASM0-2, A Fuel Assembly Burnup

Program, User's Manual,"

STUDSVIK/NR-81/3, Studsvik Energiteknik AB (March 1981).

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